ML032310200

Tech Spec Pages for Amendment No. 221, Use of a Pressure Temperature Limits Report (Ptlr), Change the Minimum Boltup Temperature, Revise the Low Temperature Overpressure Protection (LTOP) Methodology and Analysis

ML032310200
Person / Time
Site:	Fort Calhoun
Issue date:	08/15/2003
From:	NRC/NRR/DLPM/LPD4
To:
Wang A, NRR/DLPM, 415-1445
Shared Package
ml032300311	List: ML032310200
References
TAC MB6468
Download: ML032310200 (19)

Text

TABLE OF CONTENTS (Continued)

Page 4.3 Nuclear Steam Supply System (NSSS) .............................................. 4-3 4.3.1 Reactor Coolant System (RCS) ............................................ 4-3 4.3.2 Reactor Core and Control .......................... ....................... 4-3 4.3.3 Emergency Core Cooling ................... 4-3 4.4 Fuel Storage ................... 4-4 4.4.1 New Fuel Storage ................................................. ..... 4-4 44.2 Spent Fuel Storage ................... 4-4 4.5 Seismic Design for Class I Systems ................... 4-5 5.0 ADMINISTRATIVE CONTROLS . ........................................................ 5-1 5.1 Responsibility ....................................... . .................... 5-1 5.2 Organization ................... 5-2 5.3 Facility Staff Qualifications ................................... .. ............... 5-2 5.4 Training . . . 5-4 5.5 Review and Audit . . .5-4 5.5.1 Plant Review Committee (PRC) ........................................ . 5-4 5.5.2 Safety Audit and Review Committee (SARC) .. 5-4 56 Reportable Event Action ... 5-4 5.7 Safety Limit Violation ..................................................... ...... 5-5 5.8 Procedures ........... ...... .. .. ... 5-5 59 Reporting Requirements ..... ........... . ...... ... 5-6 5 9.1 Routine Reports .. ......................... . .......... ............. 5-6 5.9.2 Reportable Events ... ... ...... ....... . ..... 5-7 5 9.3 Special Reports ..... ..... ... . . ... . .... .. 5-7 5.9.4 Unique Reporting Requirements .......... .... .. .................. .... 5-8 5 9.5 Core Operating Limits Report ............. .......... ..... .. . ........ . .. 58 5.9.6 RCS Pressure-Temperature Limits Report (PTLR) .. 5-10a 5.10 Records Retention .... ..................

.. .. . . .... ... ........ ..... S11 11.

5.11 Radiation Protection Program . .. . ..... 5-11 5.12 DELETED 5.13 Secondary Water Chemistry ..... . .... ......... 5-12 5.14 Systems Integrity ... 5-12 5.15 Post-Accident Radiological Sampling and Monitoring . . . 5-13 5.16 Radiological Effluents and Environmental Monitoring Programs .. ............ 5-13 5.16.1 Radioactive Effluent Controls Program. . . . ... .... 5-13 5.16.2 Radiological Environmental Monitoring Program ... ... ..... .. ... 5-14 5.17 Offsite Dose Calculation Manual (ODCM) ........ ... ...... 5-15 5.18 Process Control Program (PCP) ..................... . . .......... S15 5..1.5..

5.19 Containment Leakage Rate Testing Program . ................ ................ 5-16 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS ..... ...... ..... .. . ....... . 6-1 6.1 DELETED 6.2 DELETED 6.3 DELETED 6.4 DELETED iii Amendment No ZZ,34,4,5A,557, 73eGO,0c.0ZDZ,,1 41,1 52,i V7,164,1B,221

TECHNICAL SPECIFICATIONS - FIGURES TABLE OF CONTENTS PAGE WHICH FIGURE DESCRIPTION FIGURE FOLLOWS 1-1 TMLP Safety Limits 4 Pump Operations ............................. 1-3 2-1A Deleted I 2-1B Deleted 2-3 Deleted 2-12 Boric Acid Solubility In Water . ............................. ............... 2-1 gh 2-10 Spent Fuel Pool Region 2 Storage Criteria ........ ..................... .. 2-39e 2-8 Flux Peaking Augmentation Factors ............................. 2-53 viii Amendment No. 11C,12C,131,1A1,1CI,170, 172, 1, 192,1^ 7, 207 21

DEFINITIONS E - Average Disintegration Energy E is the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration, in MEV, for isotopes, other than iodines, with half lives greater than 15 minutes making up at least 95% of the total non-iodine radioactivity in the coolant.

Offsite Dose Calculation Manual (ODCM)

The document(s) that contain the methodology and parameters used in the calculations of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent radiation monitoring Warn/High (trip) Alarm setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain:

1) The Radiological Effluent Controls and the Radiological Environmental Monitoring Program required by Specification 5.16.

2) Descriptions of the information that should be included in the Annual Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports required by Specifications 5.9.4.a and 5.9.4.b.

Unrestricted Area Any area at or beyond the site boundary access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

Core Operating Limits Report (COLR)

The Core Operating Limits Report (COLR) is a Fort Calhoun Station Unit No. 1 specific document that provides core operating limits for the current operating cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Section 5.9.5. Plant operation within these operating limits is addressed in the individual specifications.

RCS Pressure-Temperature Limits Report (PTLR)

The PTLR is a fluence dependent document that provides Limiting Conditions for Operation (LCO) in the form of pressure-temperature (P-T) limits to ensure prevention of brittle fracture. In addition, this document establishes power operated relief valve setpoints which provide low temperature overpressure protection (LTOP) to assure the P-T limits are not exceeded during the most limiting LTOP event. The P-T limits and LTOP criteria in the PTLR are applicable through the effective full power years (EFPYs) specified in the PTLR. NRC approved methodologies are used as the bases for the information provided in the PTLR.

References (1) USAR, Section 7.2 (2) USAR, Section 7.3 8 Amendment No. 67,86,141,52,164,221

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Operable Components (Continued)

(5) DELETEb (6) Both steam generators shall be filled above the low steam generator water level trip set point and available to remove decay heat whenever the average temperature of the reactor coolant is above 300 0F. Each steam generator shall be demonstrated operable by performance of the inservice inspection program specified in Section 3.17 prior to exceeding a reactor coolant temperature of 3000 F.

(7) Maximum reactor coolant system hydrostatic test pressure shall be 3125 psia. A maximum of 10 cycles of 3125 psia hydrostatic tests are allowed.

(8) Reactor coolant system leak and hydrostatic test shall be conducted within the limitations of the pressure and temperature limit Figure(s) shown in the PTLR.

(9) Maximum secondary hydrostatic test pressure shall not exceed 1250 psia. A minimum measured temperature of 730 F is required. Only 10 cycles are permitted.

(10) Maximum steam generator steam side leak test pressure shall not exceed 1000 psia. A minimum measured temperature of 730 F is required.

(11) Low Tmperature Overpressure Protection (LTOP)

(a) The LTOP enable temperature and RCP operations shall be maintained in accordance with the PTLR.

(b) The unit can not be placed on shutdown cooling until the RCS has cooled to an indicated RCS temperature of less than or equal to 300'F.

(c) If no reactor coolant pumps are operating, a non-operating reactor coolant pump shall not be started while T. is below the LTOP enable temperature stated in the PTLR unless there is a minimum indicated pressurizer steam space of at least 50% by volume.

2-2a Amendment No. ,S6,66,71,119,136, 161,1680-2x6, 221

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 ODerable Components (Continued)

(12) Reactor Coolant System Pressure Isolation Valves (a) The integrity of all pressure isolation valves listed in Table 2.9 shall be demonstrated, except as specified in (b). Valve leakage shall not exceed the amounts indicated.

(b) In the event that the integrity of any pressure isolation valve specified in Table 2-9 cannot be demonstrated, reactor operation may continue, provided that at least two valves in each high pressure line having a nonfunctional valve are in and remain in the mode corresponding to the isolated condition. Manual valves shall be locked in the closed position; motor operated valves shall be placed in the closed position and power supply deenergized.

(c) If Specifications (a) and (b) above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Basis The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation and maintain DNBR above 1.18 during all normal operations and anticipated transients.

When Specification 2.1.1(2) is applicable, the reactor coolant pumps (RCPs) are used to provide forced circulation heat removal during heatup and cooldown. Under these conditions, decay heat removal requirements are low enough that a single reactor coolant system (RCS) loop with one RCP is sufficient to remove core decay heat.

However, two RCS loops are required to be OPERABLE to provide redundant paths for decay heat removal. Only one RCP needs to be OPERABLE to declare the associated RCS loop OPERABLE. Reactor coolant natural circulation is not normally used but is sufficient for core cooling. However, natural circulation does not provide turbulent flow conditions. Therefore, boron reduction in natural circulation is prohibited because mixing to obtain a homogeneous concentration in all portions of the RCS cannot be assured.

2-2b Amendment No. 56,77,92, , 188,221

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Operable Components (Continued)

When Specification 2.1.1(3) is applicable, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat, but single failure considerations require that at least two loops be operable. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two shutdown cooling pumps to be OPERABLE.

One of the conditions for which Specification 2.1.1(3) is applicable is when the RCS temperature ad) is less than 210 0F, fuel is in the reactor and all reactor vessel closure bolts are fully tensioned. As soon as a reactor vessel head closure bolt is loosened, Specification 2.1.1(3) no longer applies, and Specification 2.8 is applicable.

Specification 2.8 also requires two shutdown cooling loops to be operable if there is less than 23 feet of water above the top of the core.

The restrictions on availability of the containment spray pumps for shutdown cooling service ensure that the SI/CS pumps' suction header piping is not subjected to an unanalyzed condition in this mode. Analysis has determined that the minimum required RCS vent area is 47 in2. This requirement may be met by removal of the pressurizer manway which has a cross-sectional area greater than 47 in2.

When reactor coolant boron concentration is being changed, the process must be uniform throughout the reactor coolant system volume to prevent stratification of reactor coolant at lower boron concentration which could result in a reactivity insertion. Sufficient mixing of the reactor coolant is assured if one low pressure safety injection pump or one reactor coolant pump is in operation. The low pressure safety injection pump will circulate the reactor coolant system volume in less than 35 minutes when operated at rated capacity. The pressurizer volume is relatively inactive; therefore, it will tend to have a boron concentration higher than the rest of the reactor coolant system during a dilution operation. Administrative procedures will provide for use of pressurizer sprays to maintain a nominal spread between the boron concentration in the pressurizer and the reactor coolant system during the addition of boron.0" Both steam generators are required to be filled above the low steam generator water level trip set point whenever the temperature of the reactor coolant is greater than the design temperature of the shutdown cooling system to assure a redundant heat removal system for the reactor.

The bases for the LTOP system requirements are documented in the PTLR.

2-2c Amendment No. 56 71,136,1q1, 18,221

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Operable Components (Continued)

The design cyclic transients for the reactor system are given in USAR Section 4.2.2. In addition, the steam generators are designed for additional conditions listed in USAR Section 4.3:4. Flooded and pressurized conditions on the steam side assure minimum tube sheet temperature differential during leak testing. The minimum temperature for pressurizing the steam generator steam side is 70'F; in measuring this temperature, the instrument accuracy must be added to the 700 F; limit to determine the actual measured limit. The measured temperature limit will be 73'F based upon use of an instrument with a maximum inaccuracy of +/- 2 0 F and an additional 1 F safety margin.

References (1) USAR Section 4.3.7 2-2d Amendment No. 568 *,-+e,1 ,8, 221

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatug and Cooldown Rate Aplicability Applies to the temperature change rates and pressure of the Reactor Coolant System (RCS).

Objective To specify limiting condiftions of the reactor coolant system heatup and cooldown rates.

Specification The combination of RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR and as designated below:

a. Allowable combinations of pressure and temperature (Ta) for a specific heatup rate shall be below and to the right of the applicable limit lines as shown on the pressure and temperature (P-T) limit Figure(s) in the PTLR.

b. Allowable combinations of pressure and temperature (Ta) for a specific cooldown rate shall be below and to the right of the applicable limit lines as shown on the P-T limit Figure(s) in the PTLR.

c. The heatup rate of the pressurizer shall not exceed 100 0 F in any one hour period.

d. The cooldown rate of the pressurizer shall not exceed 200OF in any one hour period.

Required Actions (1) When any of the above limits are exceeded, the following corrective actions shall be taken:

(a) Immediately initiate action to restore the temperature or pressure to within the limit.

(b) Perform an analysis to determine the effects of the out of limit condition on the fracture toughness properties of the reactor coolant system.

(c) Determine that the reactor coolant system remains acceptable for continued operation or be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

(2) Before the radiation exposure of the reactor vessel exceeds the exposure for which they apply, the P-T limit Figure(s) shown in the PTLR shall be updated in accordance with the following criteria and procedures:

2-3 Amendment No. 22,74,161, 207,221

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

(a) The P-T limit Figure(s) are valid for a fast neutron (E 1 MeV) fluence value and corresponding EFPY as stated in the PTLR.

(b) The limit line on the P-T limit Figure(s) shown in the PTLR shall be updated for a new integrated power period as follows: the total integrated reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated fast neutron exposure (Ez1 MeV).

(c) The limit lines in the P-T limit Figure(s) shown in the PTLR shall be moved parallel to the temperature axis (horizontal) in the direction of increasing temperature a distance equivalent to the transition temperature shift during the period since the curves were last constructed. The boltup temperature limit line shall remain at 640 F as it is set by the RTNDT of the reactor vessel flange and is not subjected to a fast neutron flux. The lowest service temperature shall remain at 1640 F because components related to this temperature are also not subjected to a fast neutron flux.

(d) Technical Specification 2.3(3) shall be reviewed and revised as necessary each time the curves on the P-T limit Figure(s) as shown in the PTLR are revised.

Basis All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor coolant system temperature and pressure changes.')

These cyclic loads are introduced by normal unit load transients, reactor trips and startup and shutdown operation.

During unit startup and shutdown, the rates of temperature and pressure changes are limited. The design number of cycles for heatup and cooldown is based upon allowable heatuplcooldown rates and cyclic operation. Cycle dependent information such as the pressure-temperature limit curves, low temperature overpressure protection system limits, neutron fuence, and adjusted reference temperatures are contained in the PTLR, which was developed using the methodologies stated in Technical Specification 5.9.6(b) and in the PTLR(2 ).

2-4 Amendment No. 22,47,64,74,77,100,114, 161,197, 6, 221

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 HeatuD and Cooldown Rate (Continued)

References:

(1) USAR, Section 4.2.2 (2) Technical Data Book IX, Fort Calhoun Station Unit No. 1, RCS I Pressure-Temperature Limits Report I

2-7a Amendment No. 92,47,64,74,1 O0,1 ,197,

,07, 221

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves Applicability Applies to the status of the pressurizer and main steam safety valves.

Obiective To specify minimum requirements pertaining to the pressurizer and main steam safety valves.

Specifications To provide adequate overpressure protection for the reactor coolant system and steam system, the following safety valve requirements shall be met:

(1) The reactor shall not be made critical unless the two pressurizer safety valves are operable with their lift settings adjusted to ensure valve opening at 2485 psig +/-1%

and 2530 psig +/-1 %.(')

(2) Whenever there is fuel in the reactor, and the reactor vessel head is installed, a minimum of one operable safety valve shall be installed on the pressurizer.

However, when in at least the cold shutdown condition, safety valve nozzles may be open to containment atmosphere during performance of safety valve tests or maintenance to satisfy this specification.

(3) At least four of the five Main Steam Safety Valves (MSSVs) associated with each steam generator shall be OPERABLE in MODES 1 and 2. Lift settings shall be at 985 psig +3/-2%, 1000 psig +3/-2%, 1010 psig +31-2%, 1025 psig +31-2%, and 1035 psig +3/-2%.('1

a. With less than four of the five MSSVs associated with each steam generator OPERABLE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT SHUTDOWN within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(4) Two power-operated relief valves (PORVs) shall be operable during heatups and cooldowns when the RCS temperature is less than 515°F, and in Modes 4 and 5 whenever the head is on the reactor vessel and the RCS is not vented through a 0.94 square inch or larger vent, to prevent violation of the pressure-temperature limits designated by the P-T limit Figure(s) shown in the PTLR.

a. With one PORV inoperable during heatups and cooldowns when the RCS temperature is less than 51 5 0F, restore the inoperable PORV to operable within 7 days or be in cold shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

b. With both PORVs inoperable during heatups and cooldowns when the RCS temperature is less than 515 0F, be in cold shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

c. With one PORV inoperable in Modes 4 or 5, within one hour ensure the pressurizer steam space is greater than 50% volume and restore the inoperable PORV to operable within 7 days. If adequate steam space cannot be established within one hour, then restore the inoperable PORV to operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the PORV cannot be restored in the required time, depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

2-15 Amendment No. 39,47,54,1i4,16i,18 ,267,221

2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergencv Core Cooling System Applicability Applies to the operating status of the emergency core cooling system.

Obiective To assure operability of equipment required to remove decay heat from the core.

Specifications (1) Minimum Requirements The reactor shall not be made critical unless all of the following conditions are met:

a. The SIRW tank contains not less than 283,000 gallons of water with a boron concentration of at least the refueling boron concentration at a temperature not less than 50 0F.

b. One means of temperature indication (local) of the SIRW tank is operable.

c. All four safety injection tanks are operable and pressurized to at least 240 psig and a maximum of 275 psig with a tank level of at least 116.2 inches (67%) and a maximum level of 128.1 inches (74%) with refueling boron concentration.

d. One level and one pressure instrument is operable on each safety injection tank.

e. One low-pressure safety injection train is operable on each associated 4,160 V engineered safety feature bus.

f. One high-pressure safety injection pump is operable on each associated 4,160 V engineered safety feature bus.

9. Both shutdown heat exchangers are operable.

h. Piping and valves shall be operable to provide two flow paths from the SIRW tank to the reactor coolant system.

i. All valves, piping and interlocks associated with the above components and required to function during accident conditions are operable. HCV-2914, 2934, 2974, and 2954 shall have power removed from the motor operators by locking open the circuit breakers in the power supply lines to the valve motor operators.

. FCV-326 shall be locked open.

2-20 Amendment No. 132,4,1,I1 I1 9,133,141 ,157,175j@24-,221

2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Cooling System (Continued)

(3) Protection Against Low Temperature Overgressurization The following limiting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the HPSI pumps need not be required if the RCS is vented through at least a 0.94 square inch or larger vent.

Whenever the reactor coolant system cold leg temperature is below 3500F, at least one (1) HPSI pump shall be disabled.

Whenever the reactor coolant system cold leg temperature is below 320'F, at least two (2) HPSI pumps shall be disabled.

Whenever the reactor coolant system cold leg temperature is below 270'F, all three (3) HPSI pumps shall be disabled.

Inthe event that no charging pumps are operable when the reactor coolant system cold leg temperature is below 2700F, a single HPSI pump may be made operable and utilized for boric acid injection to the core, with flow rate restricted to no greater than 120 gpm.

(4) Trisodium Phosphate (TSP) Dodecahydrate During operating Modes 1 and 2, the TSP baskets shall contain 126 ft3 of active TSP.

a. With the above TSP requirements not within limits, the TSP shall be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

b. With Specification 2.3(4)a required action and completion time not met, the plant shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Basis The normal procedure for starting the reactor is to first heat the reactor coolant to near operating temperature by running the reactor coolant pumps. The reactor is then made critical. The energy stored in the reactor coolant during the approach to criticality is substantially equal to that during power operation and therefore all engineered safety features and auxiliary cooling systems are required to be fully operable.

2-22 Amendment No. 17,39,43,47,64,74,77, 00,103,133,141,157,161,179,,01,221 July 15, 1999

2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Cooling System (Continued)

Components in excess of those allowed by Conditions a, b, d, and e may be inoperable provided they are returned to operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> when performing the quarterly recirculation actuation logic channel functional test (Table 3-2 item 20) under administrative controls. This allowance applies only to the remaining portion of Cycle 20 and all of Cycle 21. This prevents violating Technical Specifications or necessitating a unit shutdown due to inability to perform the quarterly recirculation actuation logic channel functional test. These administrative controls consist of stationing three dedicated operators at the Engineered Safeguards Features (ESF) panel controls in the control room. In this way, the following conditions are maintained and actions can be rapidly performed should a valid ESF actuation occur:

the appropriate Safety Injection Refueling Water Tank (SIRWT) to Safety Injection (SI) and Containment Spray (CS) pumps suction valve control switch is maintained in the OPEN position (spring-retum switch),

the appropriate Si and CS pumps to SIRWT recirculation minimum flow valve control switch is maintained in the OPEN position (spring-return switch),

the appropriate Recirculation Actuation Signal (RAS) lockout relays and initiating signal can be rapidly reset, the appropriate Si and CS pumps to SIRWT recirculation minimum flow valve control switch can be rapidly returned to the AUTO position, the appropriate SIRWT to Si and CS pumps suction valve control switch can be rapidly returned to the AUTO position, and the appropriate Containment Sump to SI and CS pumps suction valve control switch can be rapidly returned to the AUTO position.

The appropriate SI and CS pumps to SIRWT recirculation minimum flow valve control switch and the appropriate SIRWT to Si and CS pumps suction valve control switch are held in the OPEN position during the test to enhance the reliability of the appropriate SI and CS pumps by maintaining the associated valves open.

References (1) USAR, Section 14.15.1 (2) USAR, Section 6.2.3.1 (3) USAR, Section 14.15.3 (4) USAR, Appendix K (5) Omaha Public Power District's Submittal, December 1, 1976 (6) Deleted (7) USAR, Section 4.4.3 (8) CE NPSD-994, "CEOG Joint Applications Report for Safety Injection Tank AOT/SIT Extension," May 1995.

(9) CE NPSD-995, "CEOG Joint Applications Report for Low Pressure Safety Injection System AOT Extension," May 1995.

2-23b Amendment No. 47,64,74,179,-06,--7,221

2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core 2.10.1 Minimum Conditions for Criticality Applicability Applies to the status of the reactor coolant system during reactor criticality.

Objective To prevent unanticipated power excursions of an unsafe magnitude.

Specifications (1) The reactor shall not be made critical if the average reactor coolant temperature is below 515 0F.

(2) No more than one CEA at a time in a regulating or non-trippable group shall be exercised or withdrawn until after a steam bubble and normal water level as given in operating procedures are established in the pressurizer.

(3) Reactor coolant boron concentration shall not be reduced below that required for the Hot Shutdown Condition until after a steam bubble and normal water level are established in the pressurizer.

Basis At the beginning of each fuel cycle, the moderator temperature coefficient is expected to be slightly negative at operating temperatures with all CEA's withdrawn. However, variations in cycle core loading and the uncertainty of the calculation are such that it is possible that a slightly positive coefficient could exist.

The moderator temperature coefficient at lower temperatures will be less negative or more positive than at operating temperature. It is, therefore, prudent to restrict the operation of the reactor when reactor coolant temperatures are less than 515'F.

2-48 Amendment No. 32,109, 221

2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.1 Minimum Conditions for Criticality (Continued)

If the shutdown margin required by the Hot Shutdown Condition is maintained, there is no possibility of an accidental criticality as a result of a change of moderator temperature or a decrease of coolant pressure. Normal water level is established in the pressurizer prior to the withdrawal of CEA or the dilution of boron so as to preclude the possible overpressurization of a solid reactor coolant system.

During physics tests, special operating precautions will be taken. In addition, the strong negative effect of the Doppler coefficient would limit the magnitude of a power excursion resulting from a reduction of moderator density.

2-49 Amendment No. 3,221

TABLE 3.5 (Continued)

Test Frequency

19. Refueling Water Level Verify refueling water level is 2 23 ft. above Prior to commencing, and daily during CORE ALTERATIONS the top of the reactor vessel fange. and/or REFUELING OPERATIONS inside containment.

20. Spent Fuel Pool Level Verify spent fuel pool water level is 2 23 ft. Prior to commencing, and weekly during REFUELING above the top of irradiated fuel assemblIes seated OPERATIONS In the spent fuel pool.

in the storage racks.

21. Containment Penetrations Verify each required containment penetration is Prior to commencing, and weekly during CORE ALTERATIONS in the required status. andlor REFUELING OPERATIONS in containment.

22. Spent Fuel Assembly Verify by administrative means that Initial Prior to storing the fuel assembly in Region 2 (including Storage enrichment and bumup of the fuel assembly is in peripheral cells).

accordance with Figure 2-10.

23. P-T Limit Curve Verify RCS Pressure, RCS temperature, and This test is only required during RCS heatup and cooldown RCS heatup and cooldown rates are within operations and RCS inservice leak and hydrostatic testing.

the limits specified by the P-T limit Figure(s) While these operations are occurring, this test shall be performed shown in the PTLR. every 30 minutes.

3-20f Amendment No. 468,221

3.0 SURVEILLANCE REQUIREMENTS 3.3 Reactor Coolant System and Other Components Subject to ASME Xl Boiler & Pressure Vessel Code Inspection and Testing Surveillance Applicability Applies to in-service surveillance of primary system components and other components subject to inspection and testing according to ASME XI Boiler & Pressure Vessel Code.

Objective To ensure the integrity of the reactor coolant system and other components subject to inspection and testing according to ASME Xl Boiler & Pressure Vessel Code.

Specifications (1) Surveillance of the ASME Code Class 1, 2 and 3 systems, except the steam generator tubes inspection, should be covered by ASME Xl Boiler & Pressure Vessel Code.

a. In-service inspection of ASME Code Class 1, Class 2, and Class 3 components, including applicable supports, and in-service testing of ASME Code Class 1, Class 2, and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a (g)(6)(i).

b. Surveillance of the reactor coolant pump flywheels shall be performed as indicated in Table 3-6.

c. A surveillance program to monitor radiation-induced changes in the mechanical and impact properties of the reactor vessel materials shall be maintained in accordance with 10 CFR Part 50 Appendix H.')

Examinations results shall be used to update the PTLR.

(2) Surveillance of Reactor Coolant System Pressure Isolation Valves

a. Periodic leakage testing* on each valve listed in Table 2-9 shall be accomplished prior to entering the power operation mode every time the plant is placed in the cold shutdown

• To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

3-21 Amendment No. 46,7GiO4,i42,i57,176,

5.0 ADMINISTRATIVE CONTROLS 5.9.6 Reactor Coolant System (RCS) Pressure - Temperature Limits Report (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature overpressure protection, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for Technical Specifications 2.1.1 and 2.1.2.

b. The analytical methods used in the PTLR shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. CE NPSD-683-A, Revision 6, "Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications," April 2001.

2. WCAP-15443, Revision 0, "Fast Neutron Fluence Evaluations for the Fort Calhoun Unit I Reactor Pressure Vessel," July 2000.

3. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment Number 199 to Facility Operating License DPR40 Omaha Public Power District Fort Calhoun Station, Unit Number 1, dated June 7, 2001.

4. CEN-636, Revision 2, "Evaluation of Reactor Vessel Surveillance Data Pertinent to the Fort Calhoun Reactor Vessel Beltline Materials, dated July 2000.

5. FC06876, Revision 0, "Performance of Low Temperature Overpressure Protection System Analyses Using RELAP5: Methodology Paper."

6. FC06877, Low Temperature Overpressure Protection (LTOP) Analysis, Revision 1."

7. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment Number 207 to Facility Operating License Number DPR-40 Omaha Public Power District Fort Calhoun Station, Unit Number 1, dated April 22, 2002.

8. Letter LTR-CI-01-25, Revision 0 from Westinghouse Electric Company (S.T. Byrne) to OPPD (J. Jensen), "Assessment of Extended Beltline Limit for Fort Calhoun Station Reactor Pressure Vessel," dated December 18, 2001.

9. WCAP-1 5741, Revision 0, "Reactor Vessel Surveillance Program Withdrawal Schedule Modifications," dated September 2001.

10. Letter from NRC (A. B. Wang) to Omaha Public Power District (R. T.

Ridenoure), Fort Calhoun Station - Unit 1, Exemption from the Requirements of Appendix G. to 10 CFR Part 50 (TAC No. MB8237), dated July 30, 2003.

11. Letter from Information Systems Laboratories (William Arcieri) to OPPD (J, Jensen), 'WCA-09-2002: Transmittal of RELAP5/MOD3.2d," dated August 2, 2002,,

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period (i.e., the number of EFPY used in the P-T limitILTOP analysis) and for any revision or supplement thereto.

5-1Oa Amendment No. 221